1. Field of the Invention
The present invention relates to a Method, Process or System for processing and treating a radioactive liquid or aqueous concentrate, such as a nuclear fuel plant stream, or liquid or aqueous concentrate containing radwaste or other forms of environmental waste.
2. Background Information
It has been documented that a number of plants in North America, Asia, and Europe, particularly Eastern Europe, and in other locations around the world, have been dealing with the problem of stored radioactive concentrate fluids (or radioactive agents in solution), or historical concentrates, which have, especially in the last 20-30 years grown to great stored volumes at various plants. Therefore, radionuclide removal from nuclear power plant's liquid radwaste has become an important priority for the European Union and its member states and other countries of the world. These plants have frequently included nuclear power plants where energy obtained by nuclear fission is transformed into electricity.
An example of such a plant is the Kola NPP in the Polyarnye Zori/Murmansk Region, Russian Federation. Accumulated LRW (Liquid Radioactive Waste) at this plant had, at one point, been temporarily stored in stainless steel tanks and was to have been processed in such a way as to allow safe long-term storage, haulage and final disposal of such waste. This plan had not proven to be adequately successful. The Kola NPP (Nuclear Power Plant) had operated a system for the removal of radionuclides from evaporator concentrate decantates and salt crystalline deposits. This process had consisted of an oxidation phase and a filtration phase. In their case oxidation was achieved by ozone ejection into the liquid radwaste. However, this approach did not control temperature and pH in an ideal state to further the ozone process involved, allowing it to go up to 90 degrees F. (or about 32.22 degrees C.) where soluble ozone went to about zero solubility; and, therefore, was subject to poor utilization; where it was not absorbed into water and lost as gas. The pH was not controlled in an optimum range that both prevented boron precipitation and optimized utilization of the ozone. Filtration was applied to separate (non-soluble) radioactive oxidation products from its liquid phase, but only micro-filtration rather than ultrafiltration which allowed particulate activity smaller than micro-filtration range to pass. Cobalt, silver and iron isotopes are often found in about colloidal to about the lower end of the microfiltration range. In the past some of the equipment and method approaches used in this system had been found deficient in terms of meeting the needed performance requirements and with regard to the reliability or in terms of efficiency; and in general significant improvements to this type of process have sorely been needed to address this plant and plant areas like this.
Inventions the subject of patent publication in the past suffer from a number of disadvantages; and, in one or more ways, appear to have only tangential relationship to the present invention.
See, for example: U.S. Pat. No. 4,894,091 to Napier et al. which teaches a process for removing metals from water including the steps of prefiltering solids from the water, adjusting the pH to between about 2 and 3, reducing the amount of dissolved oxygen in the water, increasing the pH to between about 6 and 8, adding water-soluble sulfide to precipitate insoluble sulfide- and hydroxide-forming metals, adding a flocculating agent, separating precipitate-containing floc, and postfiltering the resultant solution; and where the postfiltered solution may optionally be eluted through an ion exchange resin to remove residual metal ions.
U.S. Pat. No. 7,772,451 to Enda et al. discloses what is said to be a system for chemically decontaminating radioactive material, distinguishable from the present invention in providing, in its broadest sense, for “a system for chemically decontaminating radioactive material which forms a passage for liquid to flow through, comprising: a circulation loop connected to the passage for circulating a decontamination liquid, the circulation loop comprising a decontamination agent feeder feeding the decontamination liquid that is reductive and that is an aqueous solution comprising a monocarboxylic acid (namely, “formic acid”) and a di-carboxylic acid (namely, “oxalic acid”) to the decontamination liquid; a hydrogen peroxide feeder feeding hydrogen peroxide to the decontamination liquid; an ion exchanger for separating and removing metal ions in the decontamination liquid; and an ozonizer for injecting ozone into the decontamination liquid or an oxidizer feeder feeding permanganic acid or permanganate to the decontamination liquid; and wherein the system does not contain a device for reducing trivalent iron atoms into bivalent iron atoms, and wherein any acid present in the system is an organic acid. This system, as well as that of Napier et al. just above, does not employ the present invention's process steps of Oxidation or Ozone Oxidation (I) Sorption or Powder Sorbent Isotopic Reduction (II), Solid-Liquid Separation (III), Adjustable and Configurable Ion exchange (IX) (IV), and Within Step V: Discharge of Water (Va) or Drying of resulting waste stream dissolved solids to Dry Solids (Vb).
U.S. Pat. No. 5,196,124 to Connor et al. appears to involve a method for reducing the radioactive material content of fluids withdrawn from subterranean reservoirs which employs the deposition of sorbent solids within its reservoir matrix surrounding its production well to act as an in-situ filter for dissolved radionuclides present in reservoir pore waters. Though using a form of sorption application, Connor does not facilitate this use in the same manner or staging as that set forth in the present invention. It does not employ the order of steps used or the effect so obtained by Oxidation prior to sorption; or Solid-Liquid Separation, Adjustable and Configurable ion exchange, or discharge of water or drying of waste stream dissolved solids to dry solids, all after the step of sorption. See also U.S. Pat. No. 5,728,302 to Connor; engendering similar distinctions in relation to the present invention.
U.S. Pat. No. 5,908,559 to Kreisler sets forth a METHOD FOR RECOVERING AND SEPARATING METALS FROM WASTE STREAMS. The 25 method involves steps, distinguishable from the present invention, where: pH of a waste stream is adjusted; a metal complexing agent is added; a particle growth enhancer is added; a flocculating agent is added resulting in a solution; the solution effluent is dewatered, preferably using a plate and frame press, resulting in a sludge and a supernatant; and metals are recovered from the sludge upon melting, drying and dewatering a filter cake with melting enhancers so as to permit selective removal of a fused metal-bearing concentrate for casting into ingots to be sold to primary smelters.
U.S. Pat. No. 7,282,470 to Tucker et al., though utilizing a water soluble sorbent additive, namely sorbitol or mannitol; is otherwise dissimilar to the steps of the method of the present invention.
U.S. Application No. 200910252663 of Wetherill, provides for a METHOD AND SYSTEM FOR THE REMOVAL OF AN ELEMENTAL TRACE CONTAMINANT FROM A FLUID STREAM; and includes within its steps passing a fluid stream with an elemental trace contaminant through a flow-through monolith comprising an oxidation catalyst to oxidize the elemental trace contaminant; and contacting the fluid stream comprising the oxidized trace contaminant with a sorbent free of oxidation catalyst to sorb the oxidized trace contaminant. However, it otherwise lacks the functional effect brought about by the other inclusive steps of the present invention.
In the PCT publication, W02007123436 (A1) of ALEXANDROVI et al. as inventors; the disclosure appears to disclose the use of a sorbent and the use of oxidizers such as potassium permanganate. However, this process does not employ the order sequence of the 25 present invention; nor employ Solid-Liquid Separation III, Adjustable and Configurable Ion exchange (IX) IV, or Discharge of Water (Va) or Drying of resulting waste stream dissolved solids to Dry Solids (Vb), as carried out in the present invention.
The Russian patent, RU 2122753 (C1) to Dmitriev, et al. appears to set forth elements within a process which consists in oxidative treatment of waste through ozonation in the presence of oxidation catalyst and/or radionuclide collector; solid-liquid separation and, further downstream, a liquid phase finally purified on selective sorbents. However, the order sequence and qualitative composition of the steps is dissimilar to the present invention; and Dmitriev does not employ Adjustable and Configurable Ion exchange (IX) (IV), and Within Step V: Discharge of Water (Va) or Drying of resulting waste stream dissolved solids to Dry Solids (Vb) in the same manner as the present invention; nor is clear from an absence of descriptive illustration as to the routing and nature of treatment to achieve radionuclide separation.
It will, therefore, be understood by those skilled in these technologies that a substantial and distinguishable process and system with functional and structural advantages are realized in the present invention over the past conventional technology with regard to processing, treating, packaging and chemically affecting radwaste liquid or a concentrate fluid stored or located at or in relation to a nuclear plant. It will also be appreciated that the efficiency, flexibility, adaptability of operation, diverse utility, and distinguishable functional applications of the present invention all serve as important bases for novelty of the invention, in this field of technology.